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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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DOE OSTI2012

Advanced Fuel Cycle Initiative University Fellowship Program (Final Progress Report)

2004-2011 Final Report for AFCI University Fellowship Program. The goal of this effort was to be supportive of university students and university programs - particularly those students and programs that will help to strengthen the development of nuclear-related fields. The program also supported the stability of the nuclear infrastructure and developed research partnerships that are helping to enlarge the national nuclear science technology base. In this fellowship program, the U.S. Department of Energy sought master's degree students in nuclear, mechanical, or chemical engineering, engineering/applied physics, physics, chemistry, radiochemistry, or fields of science and engineering applicable to the AFCI/Gen IV/GNEP missions in order to meet future U.S. nuclear program needs. The fellowship program identified candidates and selected full time students of high-caliber who were taking nuclear courses as part of their degree programs. The DOE Academic Program Managers encouraged fellows to pursue summer internships at national laboratories and supported the students with appropriate information so that both the fellows and the nation's nuclear energy objectives were successful.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2010

Neutronic/Thermalhydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations

The objective of this project was to test new coupling algorithms and enable efficient and scalable multi-physics simulations of advanced nuclear reactors, with considerations regarding the implementation of such algorithms in massively parallel environments. Numerical tests were carried out to verify the proposed approach and the examples included some reactor transients. The project was directly related to the Sodium Fast Reactor program element of the Generation IV Nuclear Energy Systems Initiative and the Advanced Fuel cycle Initiative, and, supported the requirement of high-fidelity simulation as a mean of achieving the goals of the presidential Global Nuclear Energy Partnership (GNEP) vision.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

AFCI-2.0 Neutron Cross Section Covariance Library

The cross section covariance library has been under development by BNL-LANL collaborative effort over the last three years. The project builds on two covariance libraries developed earlier, with considerable input from BNL and LANL. In 2006, international effort under WPEC Subgroup 26 produced BOLNA covariance library by putting together data, often preliminary, from various sources for most important materials for nuclear reactor technology. This was followed in 2007 by collaborative effort of four US national laboratories to produce covariances, often of modest quality - hence the name low-fidelity, for virtually complete set of materials included in ENDF/B-VII.0. The present project is focusing on covariances of 4-5 major reaction channels for 110 materials of importance for power reactors. The work started under Global Nuclear Energy Partnership (GNEP) in 2008, which changed to Advanced Fuel Cycle Initiative (AFCI) in 2009. With the 2011 release the name has changed to the Covariance Multigroup Matrix for Advanced Reactor Applications (COMMARA) version 2.0. The primary purpose of the library is to provide covariances for AFCI data adjustment project, which is focusing on the needs of fast advanced burner reactors. Responsibility of BNL was defined as developing covariances for structural materials and fission products, management of the library and coordination of the work; LANL responsibility was defined as covariances for light nuclei and actinides. The COMMARA-2.0 covariance library has been developed by BNL-LANL collaboration for Advanced Fuel Cycle Initiative applications over the period of three years, 2008-2010. It contains covariances for 110 materials relevant to fast reactor R&D. The library is to be used together with the ENDF/B-VII.0 central values of the latest official release of US files of evaluated neutron cross sections. COMMARA-2.0 library contains neutron cross section covariances for 12 light nuclei (coolants and moderators), 78 structural materials and fission products, and 20 actinides. Covariances are given in 33-energy groups, from 10?5 eV to 19.6 MeV, obtained by processing with LANL processing code NJOY using 1/E flux. In addition to these 110 files, the library contains 20 files with nu-bar covariances, 3 files with covariances of prompt fission neutron spectra (238,239,240-Pu), and 2 files with mu-bar covariances (23-Na, 56-Fe). Over the period of three years several working versions of the library have been released and tested by ANL and INL reactor analysts. Useful feedback has been collected allowing gradual improvements of the library. In addition, QA system was developed to check basic properties and features of the whole library, allowing visual inspection of uncertainty and correlations plots, inspection of uncertainties of integral quantities with independent databases, and dispersion of cross sections between major evaluated libraries. The COMMARA-2.0 beta version of the library was released to ANL and INL reactor analysts in October 2010. The final version, described in the present report, was released in March 2011.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Improvements to Nuclear Data and Its Uncertainties by Theoretical Modeling

This project addresses three important gaps in existing evaluated nuclear data libraries that represent a significant hindrance against highly advanced modeling and simulation capabilities for the Advanced Fuel Cycle Initiative (AFCI). This project will: Develop advanced theoretical tools to compute prompt fission neutrons and gamma-ray characteristics well beyond average spectra and multiplicity, and produce new evaluated files of U and Pu isotopes, along with some minor actinides; Perform state-of-the-art fission cross-section modeling and calculations using global and microscopic model input parameters, leading to truly predictive fission cross-sections capabilities. Consistent calculations for a suite of Pu isotopes will be performed; Implement innovative data assimilation tools, which will reflect the nuclear data evaluation process much more accurately, and lead to a new generation of uncertainty quantification files. New covariance matrices will be obtained for Pu isotopes and compared to existing ones. The deployment of a fleet of safe and efficient advanced reactors that minimize radiotoxic waste and are proliferation-resistant is a clear and ambitious goal of AFCI. While in the past the design, construction and operation of a reactor were supported through empirical trials, this new phase in nuclear energy production is expected to rely heavily on advanced modeling and simulation capabilities. To be truly successful, a program for advanced simulations of innovative reactors will have to develop advanced multi-physics capabilities, to be run on massively parallel super- computers, and to incorporate adequate and precise underlying physics. And all these areas have to be developed simultaneously to achieve those ambitious goals. Of particular interest are reliable fission cross-section uncertainty estimates (including important correlations) and evaluations of prompt fission neutrons and gamma-ray spectra and uncertainties.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2010

VISION: Verifiable Fuel Cycle Simulation Model

The nuclear fuel cycle consists of a set of complex components that work together in unison. In order to support the nuclear renaissance, it is necessary to understand the impacts of changes and timing of events in any part of the fuel cycle system. The Advanced Fuel Cycle Initiative’s systems analysis group is developing a dynamic simulation model, VISION, to capture the relationships, timing, and changes in and among the fuel cycle components to help develop an understanding of how the overall fuel cycle works. This paper is an overview of the philosophy and development strategy behind VISION. The paper includes some descriptions of the model components and some examples of how to use VISION.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

Neutron Cross Section Covariances for Structural Materials and Fission Products

We describe neutron cross section covariances for 78 structural materials and fission products produced for the new US evaluated nuclear reaction library ENDF/B-VII.1. Neutron incident energies cover full range from 10{sup -5} eV to 20 MeV and covariances are primarily provided for capture, elastic and inelastic scattering as well as (n,2n). The list of materials follows priorities defined by the Advanced Fuel Cycle Initiative, the major application being data adjustment for advanced fast reactor systems. Thus, in addition to 28 structural materials and 49 fission products, the list includes also {sup 23}Na which is important fast reactor coolant. Due to extensive amount of materials, we adopted a variety of methodologies depending on the priority of a specific material. In the resolved resonance region we primarily used resonance parameter uncertainties given in Atlas of Neutron Resonances and either applied the kernel approximation to propagate these uncertainties into cross section uncertainties or resorted to simplified estimates based on integral quantities. For several priority materials we adopted MF32 covariances produced by SAMMY at ORNL, modified by us by adding MF33 covariances to account for systematic uncertainties. In the fast neutron region we resorted to three methods. The most sophisticated was EMPIRE-KALMAN method which combines experimental data from EXFOR library with nuclear reaction modeling and least-squares fitting. The two other methods used simplified estimates, either based on the propagation of nuclear reaction model parameter uncertainties or on a dispersion analysis of central cross section values in recent evaluated data files. All covariances were subject to quality assurance procedures adopted recently by CSEWG. In addition, tools were developed to allow inspection of processed covariances and computed integral quantities, and for comparing these values to data from the Atlas and the astrophysics database KADoNiS.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

AFCI-2.0 Library of Neutron Cross Section Covariances

Neutron cross section covariance library has been under development by BNL-LANL collaborative effort over the last three years. The primary purpose of the library is to provide covariances for the Advanced Fuel Cycle Initiative (AFCI) data adjustment project, which is focusing on the needs of fast advanced burner reactors. The covariances refer to central values given in the 2006 release of the U.S. neutron evaluated library ENDF/B-VII. The preliminary version (AFCI-2.0beta) has been completed in October 2010 and made available to the users for comments. In the final 2.0 release, covariances for a few materials were updated, in particular new LANL evaluations for {sup 238,240}Pu and {sup 241}Am were adopted. BNL was responsible for covariances for structural materials and fission products, management of the library and coordination of the work, while LANL was in charge of covariances for light nuclei and for actinides.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Equilibrium cycle pin by pin transport depletion calculations with DeCART

As the Advanced Fuel Cycle Initiative (AFCI) program has matured it has become more important to utilize more advanced simulation methods. The work reported here was performed as part of the AFCI fellowship program to develop and demonstrate the capability of performing high fidelity equilibrium cycle calculations. As part of the work here, a new multi-cycle analysis capability was implemented in the DeCART code which included modifying the depletion modules to perform nuclide decay calculations, implementing an assembly shuffling pattern description, and modifying iteration schemes. During the work, stability issues were uncovered with respect to converging simultaneously the neutron flux, isotopics, and fluid density and temperature distributions in 3-D. Relaxation factors were implemented which considerably improved the stability of the convergence. To demonstrate the capability two core designs were utilized, a reference UOX core and a CORAIL core. Full core equilibrium cycle calculations were performed on both cores and the discharge isotopics were compared. From this comparison it was noted that the improved modeling capability was not drastically different in its prediction of the discharge isotopics when compared to 2-D single assembly or 2-D core models. For fissile isotopes such as U-235, Pu-239, and Pu-241 the relative differences were 1.91%, 1.88%, and 0.59%), respectively. While this difference may not seem large it translates to mass differences on the order of tens of grams per assembly, which may be significant for the purposes of accounting of special nuclear material. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Resolution of Qualification Issues for Existing Structural Materials

This report gives a detailed assessment of several key technical issues that needs resolution for the existing structural materials with emphasis on application in liquid metal reactors (LMRs), in particular, sodium cooled fast reactors. The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for the Advanced Fuel Cycle Initiative (AFCI) Reactor Campaign. The report is the second deliverable in FY09 (M2505050201) under the work package “Advanced Materials Code Qualification”. The overall objective of the Advanced Materials Code Qualification project is to evaluate the key technical requirements for the qualification of currently available and future advanced materials for application in sodium reactor systems and the resolution of issues that the U.S. Nuclear Regulatory Commission (NRC) has raised in the past on structural materials in support of the design and licensing of the LMR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. However, the development of sufficient database and qualification of these materials for application in LMRs require considerable amount of time and resources. In the meantime, the currently available materials will be used in the early development of fast reactors. Nuclear structural component designs in the U.S. comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants licensing. As the LMR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Section III Subsection NH (Class 1 Components in Elevated Temperature Service). Assessment of materials performance issues and high temperature design methodology issues pertinent to the LMR were presented in an earlier report (Natesan et al. 2008). In a subsequent report (Majumdar et al. 2009), we addressed the needs in high temperature methodologies for design of various high temperature components in sodium cooled fast reactor. The present report addresses several key technical issues for the currently available structural materials such as Type 304 and 316 austenitic stainless steels and ferritic steels such as 2.25Cr-1Mo and modified 9Cr-1Mo. The 60-year design life for the LMR presents a significant challenge to the development of database, extrapolation/prediction of long-term performance, and high temperature structural design methodology. The current Subsection NH is applicable to the design life only up to 34 years. No experimental data contain test durations of 525,000 hours, and it is impractical to conduct such long-term tests in any types of testing. So far the longest creep tests for Grade 91 and Grade 92 steels have run up to 100,000 hours. It has been noted that there is a large drop in creep rupture strength in long-term tests for these high-Cr creep-resistant steels, which may result in overestimation of the long-term creep strength and allowable stress. The report addresses in detail the need for a mechanistic understanding of the structural materials, from the standpoint of the effects of thermal aging, creep deformation, creep fracture, fatigue and creep-fatigue, creep-fatigue predictive models, fatigue and creep crack growth, and fracture toughness. Based on an in-depth assessment of the available data and mechanistic understanding, key technical issues are identified and discussed for each of the property areas. Furthermore, we have proposed viable approaches to resolve the issues and prioritized our recommendations.

Natesan, K.↗
DOE OSTI2010

THERMAL ANALYSIS OF A PROPOSED TRANSPORT CASK FOR THREE ADVANCED BURNER REACTOR USED FUEL ASSEMBLIES

Preliminary studies of used fuel generated in the US Department of Energy’s Advanced Fuel Cycle Initiative have indicated that current used fuel transport casks may be insufficient for the transportation of said fuel. This work considers transport of three 5-year-cooled oxide Advanced Burner Reactor used fuel assemblies with a burn-up of 160 MWD/kg. A transport cask designed to carry these assemblies is proposed. This design employs a 7-cm-thick lead gamma shield and a 20-cm-thick NS-4-FR composite neutron shield. The temperature profile within the cask, from its center to its exterior surface, is determined by two dimensional computational fluid dynamics simulations of conduction, convection, and radiation within the cask. Simulations are performed for a cask with a smooth external surface and various neutron shield thicknesses. Separate simulations are performed for a cask with a corrugated external surface and a neutron shield thickness that satisfies shielding constraints. Resulting temperature profiles indicate that a three-assembly cask with a smooth external surface will meet fuel cladding temperature requirements but will cause outer surface temperatures to exceed the regulatory limit. A cask with a corrugated external surface will not exceed the limits for both the fuel cladding and outer surface temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Code Qualification of Structural Materials for AFCI Advanced Recycling Reactors

This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP) and the Power Reactor Innovative Small Module (PRISM), the NRC/Advisory Committee on Reactor Safeguards (ACRS) raised numerous safety-related issues regarding elevated-temperature structural integrity criteria. Most of these issues remained unresolved today. These critical licensing reviews provide a basis for the evaluation of underlying technical issues for future advanced sodium-cooled reactors. Major materials performance issues and high temperature design methodology issues pertinent to the ARR are addressed in the report. The report is organized as follows: the ARR reference design concepts proposed by the Argonne National Laboratory and four industrial consortia were reviewed first, followed by a summary of the major code qualification and licensing issues for the ARR structural materials. The available database is presented for the ASME Code-qualified structural alloys (e.g. 304, 316 stainless steels, 2.25Cr-1Mo, and mod.9Cr-1Mo), including physical properties, tensile properties, impact properties and fracture toughness, creep, fatigue, creep-fatigue interaction, microstructural stability during long-term thermal aging, material degradation in sodium environments and effects of neutron irradiation for both base metals and weld metals. An assessment of modified versions of Type 316 SS, i.e. Type 316LN and its Japanese version, 316FR, was conducted to provide a perspective for codification of 316LN or 316FR in Subsection NH. Current status and data availability of four new advanced alloys, i.e. NF616, NF616+TMT, NF709, and HT-UPS, are also addressed to identify the R&D needs for their code qualification for ARR applications. For both conventional and new alloys, issues related to high temperature design methodology are described to address the needs for improvements for the ARR design and licensing. Assessments have shown that there are significant data gaps for the full qualification and licensing of the ARR structural materials. Development and evaluation of structural materials require a variety of experimental facilities that have been seriously degraded in the past. The availability and additional needs for the key experimental facilities are summarized at the end of the report. Detailed information covered in each Chapter is given.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2010

Accelerator and Target Technology for Accelerator Driven Transmutation and Energy Production

Since the early 1990’s, accelerator driven systems (ADS) – subcritical assemblies driven by high power proton accelerators through a spallation target which is neutronically coupled to the core – have been proposed for addressing certain missions in advanced nuclear fuel cycles. Institutes throughout the world have conducted numerous programs evaluating the role of ADS in nuclear waste transmutation and energy production. In 1995, the National Research Council (NRC) issued a report on transmutation technologies, which included an evaluation of one ADS concept that was under study at that time: a large-scale system that proposed using a ~100-MW accelerator to drive a thermal, molten salt subcritical core. The NRC recognized the numerous complexities associated with the system, including the fact that, at that time, much of the high-power accelerator technology required for that ADS system had yet to be demonstrated. Consequently, the NRC report did not look favorably upon ADS. In 1999 the US Congress directed the DOE to evaluate Accelerator Transmutation of Waste (ATW) concepts and prepare a “roadmap” to develop the technology. This roadmap identified the technical issues to be resolved, assessed the impact of ATW on high-level waste disposition, and estimated the scale and cost of deploying ATW to close the fuel cycle. It also recommended that Congress fund a $281M six-year program of trade studies and R&D on key technology issues that would support a future decision on technology demonstration. From 2000 to 2002, the DOE sponsored the Advanced Accelerator Applications (AAA) program to investigate the use of ADS in “closed” nuclear fuel cycles. In 2003 the AAA program transitioned into the Advanced Fuel Cycle Initiative, and DOE-sponsored ADS research ceased, except for the continuation on a few international collaborative efforts. A number of factors contributed to this decision, primarily the small expected growth in commercial nuclear power which “negated” the principal driver for ATW/AAA, i.e., destruction of transuranics and problematic fission products assuming that there would be essentially no new construction of commercial reactors. Outside of the USA, research into ADS for both transmutation and power generation has not only continued but accelerated. In 2001 the European Technical Working Group evaluated the state of ADS technologies and recommended the construction of an experimental ADS. In 2002 an expert group, convened by the Organization for Economic Cooperation and Development’s Nuclear Energy Agency (OECD/NEA), authored a comprehensive report entitled Accelerator Driven Systems (ADS) and Fast Reactors in Advanced Nuclear Fuel Cycles. In it, they conclude, “On the whole, the development status of accelerators is well advanced, and beam powers of up to 10 MW for cyclotrons and 100 MW for linacs now appear to be feasible. However, further development is required with respect to the beam losses and especially the beam trips to avoid fast temperature and mechanical stress transients in the reactor.” Technology demonstration is now gaining momentum with the Belgian government’s announcement of its intention to construct MYRRHA, an 85-MW prototype ADS at the Belgian Nuclear Research Centre, SCK•CEN. The government has committed to finance 40% of the construction cost, and is expending €60M over the next five years to advance the design in preparation for a construction start in 2015. In addition, there are recent indications the Chinese and Indian governments are considering construction of prototype ADS facilities of similar scale. ADS technology development programs exist in Europe, Japan, South Korea, India, China and Russia which are focused on both waste transmutation and power generation.

43 PARTICLE ACCELERATORS↗
DOE OSTI2016

Comparison of Advanced Processing of Irradiated UO{sub 2} Fuels and NpO{sub 2} Targets

Advanced processing methods for UO{sub 2}-based used nuclear fuels have been developed and demonstrated in several experimental programs at the Oak Ridge National Laboratory's Radiochemical Engineering Development Center (REDC). The Consolidated Fuel Recycle Program (1979-1986) and the Advanced Fuel Cycle Initiative (2002- 2008) were sponsored by the U.S. Department of Energy Office of Nuclear Energy (DOE/NE). Currently, DOE/NE is administering a new program for the National Aeronautics and Space Administration to produce {sup 238}Pu by irradiating NpO{sub 2}, and the irradiated targets are to be chemically processed using the same equipment and shielded facilities in the REDC that were used previously to process irradiated UO{sub 2}-based fuels. Flowsheets for processing irradiated NpO{sub 2} targets are now being developed and demonstrated. There are both similarities and differences in the processing of the two types of irradiated fuels/targets, and a comparison of these is useful for application to new types of feed material. A comparison of similarities and differences for processing UO{sub 2}- and NpO{sub 2}-based feed materials has been useful in planning for advanced processing methods and new requirements. Use of existing multistage solvent extraction equipment and shielded processing facilities at the ORNL REDC has enabled development, demonstration, and application of new and improved separations methods. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

Dynamic Simulations of Advanced Fuel Cycles

Years of performing dynamic simulations of advanced nuclear fuel cycle options provide insights into how they could work and how one might transition from the current once-through fuel cycle. This paper summarizes those insights from the context of the 2005 objectives and goals of the U.S. Advanced Fuel Cycle Initiative (AFCI). Our intent is not to compare options, assess options versus those objectives and goals, nor recommend changes to those objectives and goals. Rather, we organize what we have learned from dynamic simulations in the context of the AFCI objectives for waste management, proliferation resistance, uranium utilization, and economics. Thus, we do not merely describe “lessons learned” from dynamic simulations but attempt to answer the “so what” question by using this context. The analyses have been performed using the Verifiable Fuel Cycle Simulation of Nuclear Fuel Cycle Dynamics (VISION). We observe that the 2005 objectives and goals do not address many of the inherently dynamic discriminators among advanced fuel cycle options and transitions thereof.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Real-Time Detection Methods to Monitor TRU Compositions in UREX+Process Streams

The U.S. Department of Energy has developed advanced methods for reprocessing spent nuclear fuel. The majority of this development was accomplished under the Advanced Fuel Cycle Initiative (AFCI), building on the strong legacy of process development R&D over the past 50 years. The most prominent processing method under development is named UREX+. The name refers to a family of processing methods that begin with the Uranium Extraction (UREX) process and incorporate a variety of other methods to separate uranium, selected fission products, and the transuranic (TRU) isotopes from dissolved spent nuclear fuel. It is important to consider issues such as safeguards strategies and materials control and accountability methods. Monitoring of higher actinides during aqueous separations is a critical research area. By providing on-line materials accountability for the processes, covert diversion of the materials streams becomes much more difficult. The importance of the nuclear fuel cycle continues to rise on national and international agendas. The U.S. Department of Energy is evaluating and developing advanced methods for safeguarding nuclear materials along with instrumentation in various stages of the fuel cycle, especially in material balance areas (MBAs) and during reprocessing of used nuclear fuel. One of the challenges related to the implementation of any type of MBA and/or reprocessing technology (e.g., PUREX or UREX) is the real-time quantification and control of the transuranic (TRU) isotopes as they move through the process. Monitoring of higher actinides from their neutron emission (including multiplicity) and alpha signatures during transit in MBAs and in aqueous separations is a critical research area. By providing on-line real-time materials accountability, diversion of the materials becomes much more difficult. The objective of this consortium was to develop real time detection methods to monitor the efficacy of the UREX+ process and to safeguard the separated TRUs against unlawful diversion from within a processing facility. To achieve this, a comprehensive strategy was implemented to incorporate traditional detectors and advanced Tensioned Metastable Fluid (TMFD) metastable fluid detectors (developed, in part, under this project) into a novel detector assembly coupled to the UREX+ centrifugal contactor array. The sections below provide a brief summary of the technical achievements completed during this project. The principal outcomes are documented in more complete details contained the doctoral dissertations and masters theses, journal papers, conference proceedings and additional items for more than the 35 publications that are listed in the program bibliography in Section 3.

McDeavitt, Sean M.↗
DOE OSTI2017

Advanced Fuel Cycle Cost Basis – 2017 Edition

This report, commissioned by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), provides a comprehensive set of cost data supporting a cost analysis for the relative economic comparison of options for use in the DOE Nuclear Technology Research and Development (NTRD) Program (previously the Fuel Cycle Research and Development (FCRD) and the Advanced Fuel Cycle Initiative (AFCI)). The report describes the NTRD cost basis development process, reference information on NTRD cost modules, a procedure for estimating fuel cycle costs, economic evaluation guidelines, and a discussion on the integration of cost data into economic computer models. This report contains reference cost data for numerous fuel cycle cost modules (modules A-O) as well as cost modules for a number of reactor types (R modules). The fuel cycle cost modules were developed in the areas of natural uranium mining and milling, thorium mining and milling, conversion, enrichment, depleted uranium disposition, fuel fabrication, interim spent fuel storage, reprocessing, waste conditioning, spent nuclear fuel (SNF) packaging, long-term monitored retrievable storage, managed decay storage, recycled product storage, near surface disposal of low-level waste (LLW), geologic repository and other disposal concepts, and transportation processes for nuclear fuel, LLW, SNF, transuranic, and high-level waste. Since its inception, this report has been periodically updated. The last such internal document was published in August 2015 while the last external edition was published in December of 2009 as INL/EXT-07-12107 and is available on the Web at URL: www.inl.gov/technicalpublications/Documents/4536700.pdf. This current report (Sept 2017) is planned to be reviewed for external release, at which time it will replace the 2009 report as an external publication. This information is used in the ongoing evaluation of nuclear fuel cycles by the NE NTRD program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

Reactor-Based Management of Used Nuclear Fuel: Assessment of Major Options

This paper discusses the current status of the ongoing Advanced Fuel Cycle Initiative (AFCI) program in the U.S. Department of Energy that is investigating the potential for using the processing and recycling of used nuclear fuel to improve radioactive waste management, including used fuel. A key element of the strategies is to use nuclear reactors for further irradiation of recovered chemical elements to transmute certain long-lived highly-radioactive isotopes into less hazardous isotopes. Both thermal and fast neutron spectrum reactors are being studied as part of integrated nuclear energy systems where separations, transmutation, and disposal are considered. Radiotoxicity is being used as one of the metrics for estimating the hazard of used fuel and the processing of wastes resulting from separations and recycle-fuel fabrication. Decay heat from the used fuel and/or wastes destined for disposal is used as a metric for use of a geologic repository. Results to date indicate that the most promising options appear to be those using fast reactors in a repeated recycle mode to limit buildup of higher actinides, since the transuranic elements are a key contributor to the radiotoxicity and decay heat. Using such an approach, there could be much lower environmental impact from the high-level waste as compared to direct disposal of the used fuel, but there would likely be greater generation of low-level wastes that will also require disposal. An additional potential waste management benefit is having the ability to tailor waste forms and contents to one or more targeted disposal environments (i.e., to be able to put waste in environments best-suited for the waste contents and forms).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

History of Resistance Welding Oxide Dispersion Strengthened Cladding and other High Temperature Materials at Center for Advanced Energy Studies

Research proposal 08-1079, 'A Comparative Study of Welded ODS Cladding Materials for AFCI/GNEP,' was funded in 2008 under an Advanced Fuel Cycle Initiative (AFCI) Research and Development Funding Opportunity, number DE-PS07-08ID14906. Th proposal sought to conduct research on joining oxide dispersion strengthen (ODS) tubing material to a solid end plug. This document summarizes the scientific and technical progress achieved during the project, which ran from 2008 to 2011.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗